EP0054604B1 - Process for preparing spent solid bodies for the final disposal of radioactive wastes - Google Patents

Description

The invention relates to a process for the production of waste solidification products that are ready for disposal and contain radioactive substances with high radiation resistance or low radiolysis gas formation from the category of the radioactive waste moldings consolidated with hydraulically setting, hardening, inorganic materials, with soft radioactive, liquid, aqueous and / or radioactive solid wastes are mixed or coated with a cement-water slurry, allowed to harden and subjected to a heat treatment.

• - hoch- und mittelradioaktive Stoffe und/oder Aktiniden enthaltende, wäßrige Abfallkonzentrate,
• - in Wasser aufgeschlämmte, feinkörnige, feste Abfälle
• - aktinidenhaltigeverbrennungsrückstände,
• - feste, stückige Abfälle, wie Brennelementhülsen oder Kern-Bauteile etc.

The radioactive, liquid, aqueous and / or radioactive solid wastes, which are solidified with a high degree of safety for the environment in the present process and incorporated in the hardening shaped bodies, are those such as, for example

• aqueous and high-level radioactive substances and / or actinides containing waste concentrates,
• - fine-grained, solid waste suspended in water
• - combustion residues containing actinides,
• - solid, lumpy waste, such as fuel element sleeves or core components etc.

For the solidification of radioactive, aqueous waste, it has long been proposed to reduce the volume of such waste, thereby concentrating the radioactive substances, or subjecting the waste concentrates to heat treatment either with the addition of glass formers until the radioactive substances are distributed in the resulting glass melt and then allow the melt to solidify to a solid (DE-A-2 609 299), or mix with silicate-containing clays or ion exchangers and burn ceramic to form a solid (DE-B-2 726 087).

The disadvantages of the production of glass blocks which contain radioactive waste materials can be seen in the fact that this requires a relatively complicated and costly process technology. Another disadvantage, both in the solidification in glass, as well as in clays that are burned, is that in high-temperature steps from the not yet solidified waste significant amounts of radioactive materials, such as. As ruthenium, cesium, etc., volatilize and must be retained by a complex exhaust gas cleaning with solid filters, condensate separators and washing columns.

Cement mixtures are currently used to solidify solid, lumpy waste such as fuel element sleeves or core components. The products are not post-treated thermally. The disadvantages of cemented high-level radioactive or medium-level radioactive waste, ash, fuel element sleeves and core component products are that, due to the comparatively high water content under the influence of the radioactive radiation emanating from the waste, the radiation-chemical decomposition of the water causes large amounts of hydrogen and Oxygen are formed. This could have disadvantages for the operational safety of the intermediate and final storage. In addition, the temperatures of the products during storage must be regulated so that they do not exceed 100 ° C. At higher temperatures, water is released from the products, which leads to significant pressure build-up in the product itself and in the container. If the container is corroded, this can lead to the release of activity from the container.

The invention is based on the object of creating a process for the production of waste solidification products ready for disposal, containing radioactive substances, with high radiation resistance or less radiolysis gas formation from the category of the radioactive waste moldings solidified with hydraulically setting, hardening, inorganic materials , which all or almost all positive properties of the known solidification products also have, but the disadvantages of the known products, in particular the formation of radiolysis gas, the pressure build-up in the product or in the container, the risk of activity being released from the container or the volatilization of radioactive nuclides in the manufacture during the high temperature steps etc.

The object is achieved in a surprisingly simple manner according to the invention in that the mixtures of waste and cement-water slurry are hardened with a cement containing Portland cement clinker in closed containers at a temperature between room temperature and 150.degree. C. to give shaped bodies and in that the hardened moldings are added Normal pressure or at a negative pressure, a heat treatment at temperatures from 100 ° C to a maximum of 250 ° C to remove the pores of the molded bodies from free water.

• a) der radioaktive Abfall mit einer Aufschlämmung eines portlandzementklinkerfreien Zementes, der entweder 20 bis 30 Gew.-% Si0₂ und 40 bis 70 Gew.-% CaO oder 25 bis 35 Gew.-% Si0₂, 10 bis 25 Gew.-% CaO und 30 bis 50 Gew.-% A1₂0₃ enthält, vermischt bzw. umhüllt wird,
• b) die aus a) erhaltenen Mischungen in geschlossenen Behältern bei Raumtemperatur oder bei einer Temperaturoberhalb Raumtemperatur bis 150° Czu geformten Körpern erhärten gelassenwerden,
• c) die erhärteten Formkörper bei Normaldruck oder bei einen Unterdruck einer Wärmebehandlung bei Temperaturen von 100°C bis maximal 250°C zur Entfernung des in den Poren der Formkörper ungebunden vorliegenden Wassers unterzogen werden.

An advantageous embodiment of the invention using special elements is characterized in that

• a) the radioactive waste with a slurry of a Portland cement clinker-free cement containing either 20 to 30% by weight of SiO ₂ and 40 to 70% by weight of CaO or 25 to 35% by weight of SiO ₂ , 10 to 25% by weight CaO and 30 to 50 wt .-% A1 ₂ 0 ₃ contains, mixed or coated,
• b) the mixtures obtained from a) are allowed to harden into shaped bodies in closed containers at room temperature or at a temperature above room temperature up to 150 ° C,
• c) the hardened moldings at normal pressure or under a negative pressure of a heat treatment at temperatures of 100 ° C to a maximum of 250 ° C to remove the unbound water in the pores of the moldings.

The cement used can also contain additives such as bentonite. An advantageous embodiment of the invention is characterized in that the heat treatment of the shaped bodies (step c) is carried out at a temperature between 150 ° and 250 ° C and over a period of between 12 hours and 4 days.

• 1. Hochaktive Abfallösungen - das sind salpetersaure Lösungen von vorwiegend Schwermetallnitraten - die bei der Abtrennung der Spaltprodukte aus abgebrannten Kernbrennstoffen bei deren Wiederaufarbeitung entstehen.
• 2. Mittelaktive Abfallösungen - das sind überwiegend salpetersaure Lösungen mit in der Regel hohen Gehalten an Natriumnitrat, die an anderer Stelle als die hochaktiven Abfallösungen bei der Wiederaufarbeitung von Kernbrennstoffen und bei der Dekontamination in kerntechnischen Anlagen anfallen.
• 3. Aktinidenkonzentrate, das sind Lösungen, Pulver, oder Verbrennungsrückstände, die vor allem bei der Verarbeitung und Fabrikation von Kernbrennstoffen als Abfallprodukte erhalten werden.
• 4. Aschen und Rückstände aus der Verbrennung organischer radioaktiver Abfälle.
• 5. Feste, stückige, metallische Abfälle wie Brennelementhülsen, Kern-Bauteile, sonstige aktivierte und/oder kontaminierte metallische Abfälle, die in einem Betrieb von kerntechnischen Anlagen und bei der Wiederaufarbeitung von abgebrannten Kernbrennstoffen anfallen.

For example, in the reprocessing of nuclear fuels, in the manufacture and processing of nuclear fuels and in other nuclear facilities, the following radioactive waste is produced, which can be solidified using the method according to the invention:

• 1. Highly active waste solutions - these are nitric acid solutions of predominantly heavy metal nitrates - which arise when the fission products from spent nuclear fuels are separated and reprocessed.
• 2. Medium-active waste solutions - these are predominantly nitric acid solutions with usually high levels of sodium nitrate, which occur elsewhere than the highly active waste solutions in the reprocessing of nuclear fuels and in decontamination in nuclear facilities.
• 3. Actinide concentrates are solutions, powders, or combustion residues that are obtained as waste products, especially in the processing and manufacture of nuclear fuels.
• 4. Ashes and residues from the incineration of organic radioactive waste.
• 5. Solid, lumpy, metallic waste such as fuel element sleeves, core components, other activated and / or contaminated metallic waste that is generated in the operation of nuclear facilities and in the reprocessing of spent nuclear fuel.

The resulting aqueous solutions or aqueous slurries are mixed directly with the cement according to the invention with stirring in a container, which may also be the final storage container, or mixed indirectly in a mixing vessel and then filled into the container. The solid, lumpy waste is poured over with a mixture made of cement according to the invention and water (W / Z value advantageously 0.3 to 0.6), possibly using low pressures. The hydration (process step b)) is advantageously carried out at temperatures between room temperature and 100 ° C. The removal of the excess water (process step c)) is advantageously carried out at temperatures between 100 ° C and 200 ° C. The temperature and the duration of the heat treatment are strongly dependent on the composition and dimensions of the products. The products should be heated up relatively slowly to avoid cracking. The optimal heating and drying times must be adapted to the respective product composition and size. Heating rates in the range between 0.5 ° C./min and 2 ° C./min have proven to be advantageous.

Compared to the previous known methods for solidifying the waste solutions or slurries, such as. B. solidification in a glass matrix, etc., the inventive method has a number of significant advantages.

For example, the process technology for producing the waste products is very simple compared to glazing (mixing in the container by stirring in a separate mixing vessel and filling into the container). The hydration of the products and the removal of the excess water (of the water present in the pores of the solidification product) takes place at comparatively low temperatures, and therefore no radioactive elements escaping via the gas phase, such as e.g. B. cesium and ruthenium.

The low temperatures during product manufacture keep the corrosion rates of the equipment used very low.

By reducing the water content in the products through the thermal aftertreatment, the formation of radiolysis gas in the products is significantly reduced. The mechanical resistance is not affected by this treatment, since it is only heated to a comparatively low temperature and only excess water is separated, which is not involved in the structure of the cement stone.

By lowering the water content, the products can be stored at temperatures that are above 100 ° C to 150 ° C, since there is no longer any free water and therefore no increased water vapor partial pressure can build up and no activity can be released.

When using special cements (e.g. deep drilling cements), products are obtained which are considerably more resistant to the attack of aqueous salt solutions and are also more thermally resistant than products which correspond to the prior art.

• Durch die thermische Nachbehandlung bis ca. 250° C wird die thermodynamisch mögliche Reaktion zwischen Zirkon und Wasser schnell ablaufen und durch die entstehende Oxidschicht schnell zum Stillstand kommen. Aufgrund der thermischen Nachbehandlung steht dann kein reaktionsfähiges Wasser mehr zur Verfügung. Dies bedeutet, daß die bei Untersuchungen festgestellten hohen Tritiumfreisetzungsraten aus konventionellen Brennelementhülsen-Zementprodukten bei den Verfestigungs- produkten gemäß der Erfindung nich mehr auftreten werden. Ähnliches gilt für zementierte, tritiumhaltige Kern-Bauteile.

There is a further advantage with fuel bundle solidification products:

• Thanks to the thermal aftertreatment up to approx. 250 ° C, the thermodynamically possible reaction between zircon and water will take place quickly and will quickly come to a standstill due to the oxide layer. Due to the thermal aftertreatment there is no reactive Water more available. This means that the high tritium release rates found in investigations from conventional fuel element cement products will no longer occur in the solidification products according to the invention. The same applies to cemented, tritium-containing core components.

The invention is explained in more detail in the following examples without, however, being restricted to these examples.

example 1 Reduction of radiolys gas formation as a result of the heat treatment

• a) mit Leitungswasser (Produktbezeichnung PZ 35/Zementstein [ohne r. a. Abfall])
• b) mit simulierter mittelradioaktiver Abfallösung (Produktbezeichnung PZ 35/MAW)
• c) mit simulierter hochaktiver Abfallösung (PZ 35/HAW).

Simulated consolidation blocks on a laboratory scale (approx. 85 mm high and approx. 62 mm diameter) were produced from Portland cement 35 (PZ35) with a water / cement value W / Z = 0.4

• a) with tap water (product designation PZ 35 / cement stone [without ra waste])
• b) with simulated medium-radioactive waste solution (product designation PZ 35 / MAW)
• c) with simulated highly active waste solution (PZ 35 / HAW).

Half of the samples from a), b) and c) were produced in accordance with the prior art, ie without heat treatment, only stored for 28 days at room temperature. The other half of the samples were heat-treated by the method according to the invention, ie dewatered at 200 ° C. and normal pressure. Then all samples were irradiated with 10 MeV electrons up to 10 ⁸ rad and the gas yield measured in ml / g and Mrad. The weight loss of the dewatered samples was 23% by weight both in the samples under a) and in the samples under b) and under c). This corresponds to a residual water content of approx. 7 to 9% by weight in the product. Radiolysis gas formation for all samples is shown in the table below.

It is clear from the measured values that the samples treated by the method according to the invention had less hydrogen generation by one to two powers of ten. The table below shows the compressive strength values for the sample blocks.

Compressive strength:

The results show that the mechanical properties of the solidification products produced by the process according to the invention are not or only insignificantly deteriorated compared to the blocks produced according to the prior art.

Example 2 Reduction of radiolysis gas formation as a result of using the heat treatment a special cement

As in Example 1, laboratory blocks were produced, but using a deep drilling cement, which has no Portland cement clinker content and is to be included in the group of cements, the main components of which are in the ranges 20 to 30% by weight SiO ₂ and 40 to 70% by weight CaO fall, instead of the PZ 35. Only samples corresponding to a) in Example 1 were produced. The water-cement ratio was also 0.4, the type of irradiation and duration was the same as in Example 1. The hydrogen production rate was at 28 days stored at room temperature, the sample (prior art) 2.5 - 10- ³ ml / g and Mrad and dewatered in the group treated by the inventive process (at 200 ° C and atmospheric pressure sample only 3.7. 10- ⁴ ml / g and Mrad.

Example 3

Comparative investigations on cement stone samples without simulated waste [a) Radiolysis hydrogen yield as a function of the pre-storage time with y-irradiation; b) changes in weight when storing unirradiated samples in quinary solution at 100 ° C and

a bar]

• a) In order to simulate an enveloping solidification of solid, lumpy radioactive waste such as fuel element sleeves, which result from the reprocessing of irradiated nuclear fuel and / or fuel elements, or like core components, a cement-water mixture was made from a special cement that does not contain any Portland cement clinker and to be counted in the group of cements, the main components of which fall into the ranges 25 to 35% by weight of SiO ₂ , 10 to 25% by weight of CaO and 30 to 50% by weight of Al ₂ O ₃ , produced and used Poured into a mold at low pressure and left to harden without the addition of inactive metallic pieces. Such a simulation is permissible because the metallic pieces do not participate in the chemical process during the hardening of the cement-water mixture. Additions to metal pieces would only improve the strength properties, which, however, contribute practically nothing to the solution of the present task. The cement-water mixture had a water-cement value of W / Z = 0.45, the test specimen was cured at 100 ° C. for 4 days and then dried in a drying cabinet at normal pressure at 100 ° C. for 2 days. No cracking was then observed, the weight loss was approximately 10%. Samples produced in this way were then irradiated with y-rays and the hydrogen yield of the irradiated cement block samples produced by radiolysis was measured. For comparison, cement stone samples were made from Portland cement 35 and examined under the same conditions. The results of these tests are listed in the following tables and compared for comparison.
  
• b) Laboratory blocks, which were produced using a special cement (as described above), were found to be subject to corrosion tests during storage in quinary solution (saline solution with the composition in% by weight: 24.7% MgC1 ₂ ; 2.3% MgS0 ₄ ; 1.9% NaCI; 3.3% KCI; rest H ₂ 0) much more resistant than the blocks from conventional Portland cement, like a visual assessment on the one hand and also clearly the changes in mass of the unirradiated samples during storage in the quinary solution.
  

In Table 2, M means the weight weighed for certain samples after a certain time interval, 4M the weight change in percent of the starting weight. From the table it can easily be seen that after a certain period of time the weight change of the sample with the special cement becomes practically zero, i. H. after taking up a certain amount of the quinary solution, the mass of the sample remains the same. In contrast, the Portland cement sample (which corresponds to the prior art) swells, exhibits a significant increase in mass during increasing storage in the quinary solution and is therefore exposed to increased corrosion.

Example 4 Comparative studies on changes in weight during storage of Cement / MAW samples in quinary solution

When aqueous concentrates or slurries (simulated MAW) solidify, the concentrate is mixed with the cement. Using a special cement without PZ clinker (as described in Example 3), samples with a salt content of 10% by weight were produced with MAW simulate (W / Z = 0.45). In corrosion tests in quinary solution at 100 ° C, these products proved to be significantly more corrosion-resistant than comparable products made from conventional Portland cement.

Claims (3)

1. Method for producing waste solidification products suitable for final storage and containing radioactive substances with a high radiation resistance and a low radiolytic gas generation, respectively, of the category of the molded radioactive waste bodies which are solidified with hydraulically binding, hardenable, inorganic materials, wherein radioactive, liquid, aqueous and/or radioactive solid wastes are mixed or coated with a cement-water suspension, are allowed to harden and are subjected to a heat treatment, comprising

allowing hardening the mixtures of waste and cement-water suspension with a Portland cement klinker containing cement in closed containers at a temperature between room temperature up to 150° C to become molded bodies and

subjecting the hardened molded bodies to a heat treatment at normal pressure or negative pressure at temperatures of 100°C up to 250°C at the maximum so as to remove the water still unbound in the pores of the molded bodies.

2. Process for producing waste solidification products suitable for final storage and containing radioactive substances with a high radiation resistance and low radiolytic gas generation, respectively, of the category of molded radioactive waste bodies which are solidified with hydraulically binding, hardenable inorganic materials, wherein radioactive, liquid, aqueous and/or radioactive solid wastes are mixed or coated with a cement-water suspension, allowed to harden and subjected to a heat treatment, comprising

a) mixing and coating, respectively, the radioactive waste with a suspension of a Portland cement klinker free cement containing either 20 to 30 wt.% Si0₂ and 40 to 70 wt.% CaO or 25 to 35 wt.% SiO₂, 10 to 25 wt.% CaO and 30 to 50 wt.% Al₂O₃;

b) allowing hardening of the mixtures from a) in closed containers at room temperature or at a temperature above room temperature up to 150° C to become molded bodies;

c) subjecting the hardened molded bodies to a thermal treatment at normal pressure or at negative pressure and at temperatures of 100°C to 250°C at the maximum so as to remove the water still unbound in the pores of the molded bodies.

3. Method as described in Claim 1 or 2 wherein the heat treatment of the molded bodies (step c) is performed at a temperature between 150° C and 250° C over a period between 12 hours and four days.